The present invention relates generally to porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to methods for fabricating same.
Exploring the solar system while maintaining reasonable interplanetary travel times will require increases in spacecraft velocities of one to two orders of magnitude over currently achievable levels. The “packaged energy” for space propulsion systems required to achieve this goal, expressed as a combination of specific impulse (Isp) and mass fraction, must increase by an order of magnitude over current chemical rocket propulsion. The primary problem is that even advanced chemical rocket systems limit maneuverability and destinations. Chemical systems will not be able to extend human space exploration much beyond the moon or Mars. One solution to this problem is to utilize more energetic fuels. Therefore, NASA's goal is to develop fission propulsion to enable rapid and affordable access to any point in the solar system.
Fission power enables a new propulsion growth path with two options: nuclear thermal propulsion (NTP), in which gas is heated and expanded through a nozzle; or nuclear electric propulsion (NEP), in which nuclear power is converted to electric power for an advanced propulsion system, such as a plasma thruster. The Isp for NTP systems can increase by a factor of 10 to 100, and the specific energy can improve by a factor of 1,000,000 over conventional rockets. NTP systems have high thrust with moderate Isp (>800 sec), whereas NEP systems have lower thrust with extended Isp (>5000 sec) while providing auxiliary payload power. When compared to chemical power, fission power provides more energy for instruments, communications and higher data processing rates; more time for extended exploration; and more adaptability to changing mission requirements.
Recent conceptual designs, e.g., SAFE-100 and SAFE-300, are based on using uranium oxide (UO2) nuclear fuel elements. A 30-kW resistively heated prototype module, SAFE-30, underwent testing to verify heat transfer through an annular core geometry utilizing heat pipes and a 350-W Stirling engine power converter. Such designs clearly have potential for lower thrust, long-Isp NEP systems, but are limited by their maximum operating temperature when high-thrust, shorter-pulse NTP systems or bimodal NTP/NEP systems would best meet mission requirements.
The efficiency of the rocket increases if the temperature difference ΔT between the fuel and propellant is minimized. This dramatically improves thrust-to-mass ratio, reduces the amount of propellant required (tankage) for NTP systems, and can improve total available electric power for NEP systems when using a high-efficiency Brayton thermal cycle or advanced Stirling engine. Current designs for high-temperature gas-cooled reactor fuel elements, such as annual rods or pebble beds, cannot operate at extremely high temperatures and, thus, have limited efficiency.
U.S. Pat. No. 5,094,804 to Schweitzer and U.S. Pat. No. 4,659,911 to Bingham et al. are representative of older types of nuclear fuel elements.
Helium gas-cooled reactors have been operated in the USA using pebble-bed core designs with spherical fuel elements (60 mm diameter “pebbles”) made of TRISO fuel particles embedded in a graphite matrix. Each TRISO fuel particle is a microsphere (0.9 mm diameter) comprising a kernal/core of fissile material (UO2 or UC2) coated by multiple layers of protective barrier materials, e.g., a porous carbon buffer layer, followed by pyrocarbon, followed by silicon carbide, followed by an external coating of pyrocarbon. A typical reactor core might contain 11,000 fuel pebbles.
A new project, Prometheus, will enable advanced human exploration of the solar system, including Mars, and beyond. It specifically calls for the development of new nuclear fuels and components that are capable of extremely high temperature and very efficient operation (>925 sec Isp for >1 hr). An attractive approach would be to use high-efficiency, gas-permeable, porous nuclear fuel elements for high-temperature, gas-cooled (e.g., hydrogen, helium) space reactors. The porous fuel elements can operate at extremely high temperatures when using refractory materials with low neutron absorption cross-sections, and can have much higher heat transfer coefficients due to their very large surface-to-volume ratio and extended surface areas for transferring heat, which allows for greater power densities. The interior surfaces of the porous body could be coated with a thin layer of the nuclear fuel, so that the (hot) hydrogen or helium gas would easily flow through the interconnected open porosity to efficiently exchange heat generated by the nuclear fuel to the gas coolant. Such a porous nuclear fuel element incorporating an enriched uranium (or plutonium, thorium, americium) bi- or tri-carbide fuel could have extremely high surface area and stiffness, low density (light-weight), extremely high melting point and excellent thermal conductivity; would not degrade in hydrogen at 3000 K, would not clog; and would retain its structural integrity at high temperatures. The dispersed fissile material could be an integral part of a high-conductivity matrix. Thus, a greatly reduced temperature difference between the center of the fuel and the gas coolant (i.e., propellant) temperature could exist, which would allow the gas temperature to be much higher, yielding the high specific impulse required to sustain interplanetary exploration.
Highly porous (e.g., 90% porous) metal carbide foam structures (without nuclear fuel) have been successfully fabricated in the aerospace industry by Ultramet, Inc. through chemical vapor deposition of one or more layers of a refractory metal carbide, for example, ZrC or NbC, on a porous foam skeleton made of, for example, reticulated vitreous carbon (RVC). These metal carbide foams have been used as thermal protection systems, actively cooled structures/heat exchangers, flash and blast suppressors, and lightweight mirror substrates. The interconnected open cell geometry and tortuous flow path provides excellent heat exchange properties, excellent particulate filtration, with a correspondingly low mass. Just about any material that can be deposited by CVD/CVI techniques can be used to make a porous structure by depositing them onto a skeletal structure (e.g., RVC foam). Examples of suitable materials that can be deposited by CVD/CVI include, but are not limited to: Zr, Nb, Mo, Hf, Ta, W, Re, TiC, TaC, ZrC, SiC, HfC, BeC2, B4C, NbC, GdC, HfB2, ZrB2, Si3N4, TiO2, BeO, SiO2, ZrO2, HfO2, Y2O3, Al2O3, Sc2O3, and Ta2O5. Foam structures made of NbC and/or ZrC deposited on a RVC matrix) have several important advantages over bulk high temperature materials, such as low overall density, lack of degradation in hot hydrogen at 2700 C (where they also retain structural integrity), and minimal neutron cross-section (i.e., reduced parasitic neutron absorbtion).
Solid solution, mixed carbide fuels, such as uranium carbide (UC, UC2) and uranium bi-carbide fuels (U,Zr)C were studied in the 1970's for nuclear thermal propulsion of spacecraft as part of the Rover/NERVA program in the USA, and in similar programs in the former Soviet Union. Fuel elements designs included dispersions of small particles of UO2 or UC2 in solid graphite blocks; and a composite design made of solid-solution (U,Zr)C dispersed inside of graphite. Both fuel types were protected by a NbC or ZrC fission product barrier coating. These protective coatings were needed to protect against unacceptable mass loss due to the high chemical reactivity of free carbon with the flowing hot hydrogen propellant; and due to mis-matches in thermal expansion coefficient between the graphite matrix and the NbC or ZrC coatings. At the time the program was cancelled in 1973, there had been an evolution in thinking towards considering an all-carbide, solid-solution uranium bi-carbide fuel, e.g., (U,Zr)C or (U,Nb)C, because of their expected high resistance to erosion from exposure to hot hydrogen gas. However, the uranium bi-carbide fuels were never infiltrated into a porous matrix structure to make a porous fuel element. In addition, single-phase, solid-solution uranium tri-carbide fuels, such as (U,Zr,Nb)C, were never seriously considered because they had not been synthesized or fabricated at that point in time. It has only been in recent years that uranium tri-carbide fuels have been successfully fabricated, and their basic properties measured.
Advanced uranium (or plutonium) tri-carbide fuels have been proposed for nuclear thermal propulsion (NTP) applications because of their expected longer life and higher operating temperature; due to their high melting temperature, high thermal conductivity, and improved resistance to hot hydrogen corrosion. Recently, high density (e.g., 95%), solid solution mixed uranium/refractory metal tri-carbide fuels have been manufactured using a high-temperature liquid-phase sintering technique developed at the University of Florida (see T. W. Knight & A. Anghaie, “Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering”, Journal of Nuclear Materials 306 (2002) p. 54-60.)
These uranium tri-carbide fuels, such as (U, Zr, X) C with X=Nb, Ta, Hf, or W, exhibit high melting temperatures (greater than 3400 C) for uranium metal mole fractions of 10% or less. This melting point is almost 1000 C higher than pure UC; and UC2 has an even lower melting point. Also, UC2 erodes much more quickly in hot hydrogen than UC. Mixed phases of UC and UC2 lead to eutectic melting at a temperature 500 C lower than UC. For the tri-carbide fuels, uranium fractions greater than 10% lower the melting temperature and lead to greater uranium mass loss from either interactions with the flowing hot hydrogen propellant or vaporization from the fuel element surface, especially near the bottom of the reactor core where fuel surface and hydrogen exit temperatures might be expected to exceed 2500 C. The high solid-phase solubility of UC with the refractory metal carbides (ZrC, NbC, TaC, etc.) permits a large degree of flexibility in designing uranium bi-carbide or tri-carbide nuclear fuel elements. The carbon-to-metal (C/M) ratio can be less than 0.95 in order to maintain high melting point. When the uranium concentration is less than about 10% mole fraction, then the melting point of the tri-carbide can be as high as 3400 C for a solid solution.
Hydrogen testing on these solid solution uranium tri-carbide alloy fuels has been performed, with little erosion observed at 2700-2800 C. Additionally, the thermal conductivities of these solid-solution uranium tri-carbide fuels is much higher than conventional uranium oxide nuclear fuels (by factors of 10-20 times higher). These superior physical properties improves overall reactor efficiency and reduces system cost by allowing higher operating temperatures, reducing the amount of nuclear fuel (owing to small, more compact cores), and reducing propellant requirements (including reduced refrigeration costs). Hence, solid solution uranium tri-carbide fuels are preferred over uranium single-carbide and bi-carbide fuels.
Unfortunately, the method of fabrication described above for producing these advanced uranium tri-carbide fuels (i.e., liquid-phase sintering), cannot be used to deposit thin coatings of nuclear fuel onto exposed interior surfaces of a highly porous foam skeleton made of, e.g., reticulated vitreous or glassy carbon. The sintering step would quickly destroy the thin, interconnected structural ligaments. Therefore, the previous process of liquid-phase sintering cannot be used to make a porous, gas-permeable fuel element made of uranium tri-carbide fuel.
What is needed, then, is a high-efficiency, gas-permeable, porous nuclear fuel element for use in high temperature gas-cooled nuclear reactors (HTGR's), and a process for fabricating them, which utilizes advanced uranium bi-carbide, uranium tri-carbide, and uranium carbonitride nuclear fuels having higher melting temperatures, higher thermal conductivity, and improved resistance to corrosion from hot hydrogen gas.
Against this background, the present invention was developed.